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Hosemann, Peter

Publications and source records attributed to Hosemann, Peter.

54 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Insights on the corrosion thermodynamics of chromium in molten LiF-NaF-KF eutectic salts

The thermodynamic stability of chromium in the Cr 0 , Cr 2+ , and Cr 3+ oxidation states considering various F - ion coordination compounds was investigated in molten LiF–NaF-KF (FLiNaK) eutectic salt. Potential fluoride ion activity (F - and CrF 3 - ) diagrams were constructed to predict the most stable Cr oxidation states as a function of anion activity, solvation state of chromium ions and potential at 600 °C. The Gibbs free energies of these compounds were estimated by utilizing a combined cyclic voltammetry – Nernst theory analysis approach in FLiNaK salt. To verify the constructed diagrams, X-ray diffraction was utilized after exposure of Cr at various applied potentials to determine whether compounds detected in solidified FLiNaK salts agreed with thermodynamic calculations. This work aims to identify key thermodynamic factors that are significant to chromium corrosion for molten salt nuclear reactor applications. The F - stability region spans the entire region where spontaneous corrosion of Cr occurs. Cr may be oxidized to Cr 2+ and Cr 3+ in the presence HF (due to moisture as an impurity) spontaneously except some conditions of p 1/2 H 2 /a HF etc. This situation is not changed qualitatively for various states of solvation between oxidized Cr solute in F - solvent and this aspect is substantially similar for the two cases (pair 1: Cr 0 /CrF 3 - / CrF 6 3- ; pair 2: Cr 0 /CrF 4 2- /CrF 5 2- ) considered in this work.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Design considerations for high entropy alloys in advanced nuclear applications

The demanding operating environments of advanced nuclear reactors require the development of new nuclear materials that can withstand their increased physical, chemical, thermal, and radiation-related challenges. High-entropy alloys (HEAs) have shown often-impressive mechanical, thermomechanical, and corrosion-resistant properties, and offer a massive, unexplored compositional space which allows for the targeted development of application-specific materials. Furthermore, although still in a nascent stage, research has shown that HEAs may exhibit unique irradiation tolerance, including reduced defect production and resistance to irradiation-induced swelling and hardening. The mechanisms behind this increased tolerance are not yet well-understood, although the HEA-specific attributes of a complex energy landscape, reduced thermal conductivity, and shift in defect migration energies and pathways provide promising explanations. Here this work assesses the current and future challenges faced by structural nuclear materials, identifying the specific applications in which HEAs may provide a competitive advantage compared to industry-standard materials with the aid of Ashby material selection maps. Considerations are provided for the design of future nuclear HEAs, including calculations of nuclear-relevant properties to assist in the initial down-selection of elements depending on application requirements (e.g., low neutron capture for in-core applications), narrowing the existing compositional space of HEAs to a manageable scope.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Investigation of the fatigue crack behavior of 304 stainless steels using synchrotron X-ray tomography and diffraction: Influence of the martensite fraction and role of inclusions

Here, the effect of fatigue on the microstructure of four-point bend specimens of three variations of 304 stainless steels (Commercial 304, 304H, and 304 L) was investigated using synchrotron x-ray tomography and diffraction. X-ray tomography revealed the formation of the fatigue-induced microvoids and crack while the diffraction data was used to quantify the amount of deformation-induced martensite found after fatigue in all samples. Transmission electron microscopy evidenced the role of the precipitates/inclusions on the microvoid formation. It was found to depend on their chemical nature. The shape of the precipitates/inclusions was also found to have an effect on the microvoid shape.

304 stainless steel↗
DOE OSTI2022

The mechanism behind the high radiation tolerance of Fe-Cr alloys

With the great demand for high radiation tolerant materials for advanced nuclear energy technologies, Fe-Cr alloys are at the forefront with long standing validated performance. Yet, the real mechanism behind their high radiation resistance is in question and understanding the effect of varying Cr percentage is a grand challenge limiting further improvements. Here we applied depth resolved atomic scale probe of defects to uncover the real mechanism on how Cr improves radiation resistance and explain the controversial impact of increasing Cr percentage. By combining depth-resolved positron annihilation lifetime spectroscopy and Doppler broadening spectroscopy we investigated the effect of Cr alloying on the formation and evolution of atomic size clusters induced by ion irradiation in Fe. We also used atom probe tomography to investigate the possible presence of Cr clusters or a’ phase with high Cr composition. The study reveals that the well-known resistance to radiation in Fe-Cr alloys arises from the stabilization of vacancy clusters around Cr atoms which act as sinks for radiation-induced defects. Thus, Cr atoms do not provide a direct sink for interstitials; rather defect complexes for that consist of Cr atoms and vacancies in turn act as sinks for irradiation-induced vacancies and interstitials. Most importantly, we find that lower amounts of Cr create smaller, uniformly distributed defect clusters that act as efficient sinks for radiation damage, but larger quantities of Cr form a defect structure that is less homogenous in size and spatial distribution, resulting in less efficient damage recombination. No evidence of phase a’ was found before or after irradiation, which indicates that it does not contribute to the observed radiation tolerance.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Defect Characterization Using Positron Annihilation Spectroscopy on Laser-Ablated Surfaces

In recent years, short, pulsed laser ablation has been gaining popularity for machining small-scale test geometries from bulk samples and for efficient serial sectioning. These laser-based techniques are being added to the toolbox in material science, which makes it necessary to understand the changes in the material that occur from the laser–material interaction. Positron annihilation spectroscopy is a unique, nondestructive technique to investigate small defects in materials difficult to investigate by other tools. In this work, Doppler broadening and positron lifetime annihilation spectroscopy are utilized to help quantify the damage in materials treated with short, pulsed lasers. Using a femtosecond laser on single crystal silicon, this manuscript shows that clusters of vacancy-like defects and small voids increase systematically with laser power. Lastly, the damage induced by the laser can also reach to micrometer depths.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Commissioning of a Cask Enabling Characterization of Irradiated Nuclear Fuels with Pulsed Neutrons

Irradiation tests are a key component of nuclear fuel development and identifying typical and atypical regions in the irradiated fuel volume relies on very few characterization techniques. The goal of the effort reported here is to provide complementary measurements adding to the available parameter space for post irradiation examination as well as to inform subsequent hot cell PIE examinations by identifying typical and atypical regions with respect to microstructure, tomographic data, or isotope densities. Pulsed neutrons, enabling diffraction as well as energy-resolved neutron imaging and neutron absorption resonance spectroscopy, offer unique capabilities for this purpose. Time-of-flight neutron diffraction has the potential to offer efficient, non-destructive and non-contact microstructural characterization of irradiated fuel pins with spatial resolution of 1 mm 3 to 1 cm 3 while for energy-resolved neutron imaging (and by extension tomography) a resolution of 100 μm was demonstrated. The potential results include crystallinity vs. amorphous volumes and microstructural information such as phase compositions, lattice strains (indicative of residual stresses or chemistry variations) and textures from the diffraction data as well as distances (e.g. pellet to cladding), cracks, and isotopic distributions of minor actinides, fission products as well as fission gas partial pressures e.g. in the plenum from energy resolved neutron imaging.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Effects of Radiation-Induced Defects on Corrosion

The next generation of nuclear reactors will expose materials to conditions that, in some cases, are even more extreme than those in current fission reactors, inevitably leading to new materials science challenges. Radiation-induced damage and corrosion are two key phenomena that must be understood both independently and synergistically, but their interactions are often convoluted. In the light water reactor community, a tremendous amount of work has been done to illuminate irradiation-corrosion effects, and similar efforts are under way for heavy liquid metal and molten salt environments. While certain effects, such as radiolysis and irradiation-assisted stress corrosion cracking, are reasonably well established, the basic science of how irradiation-induced defects in the base material and the corrosion layer influence the corrosion process still presents many unanswered questions. In this review, we summarize the work that has been done to understand these coupled extremes, highlight the complex nature of this problem, and identify key knowledge gaps.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Micro-cantilever beam experiments and modeling in porous polycrystalline UO 2

Understanding the impact of microstructure on the thermo-mechanical behavior of oxide nuclear fuels is vital to predicting their performance through multiscale models. Evaluating the mechanical properties at the sub-grain length scale is key to developing these multiscale models. In this work, 3D finite element (FE) models were constructed to simulate the micrometer-scale bending of micro-cantilever beams fabricated using porous polycrystalline uranium dioxide (UO 2 ) and tested at room temperature. Here, the results showed that the porosity and elastic anisotropy of individual grains can play a significant role in determining the effective mechanical properties of the material deduced from the tests. Specifically, the porosity had a non-negligible effect, given that the pore size was of the same order of magnitude as the dimensions of the micro-beams. Correlations between load-deflection data, pore location, and elastic properties (effective Young's modulus) were investigated using UO 2 micro-beam FE models, where pore clusters were included and placed at different locations along the length of the beam. Results indicated that the presence of pore clusters near the substrate, i.e., the clamp of the micro-cantilever beam, has the strongest effect on the load-deflection behavior, with the porosity leading to a reduction of stiffness that is the largest for any location of the pore clusters. Furthermore, it was also found that pore clusters located towards the middle of the span and close to the end of the beam have a comparatively small effect on the load-deflection behavior. Therefore, it is concluded that accurate estimates of Young's modulus can be obtained from micro-cantilever experiments after accounting for porosity on the one third of the beam length close to the clamp. This, in turn, provides an avenue to improve microscale experiments and their analysis in porous, anisotropic elastic materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Iron-rich microstructure records of high temperature multi-component silicate melt behavior in nuclear fallout

Above-ground nuclear explosions that interact with the surface of the earth entrain materials from the surrounding environment, influencing the resulting physical and chemical evolution of the fireball, which can affect the final chemical phase and mobility of hazardous radionuclides that are dispersed in the environment as fallout particles. The interaction of iron with a nuclear explosion is of specific interest due to the potential for iron to act as a redox buffer and because of the likelihood of significant masses of metals to be present in urban environments. We investigated fallout from a historic surface interacting nuclear explosion conducted on a steel tower and report the discovery of widespread and diverse iron-rich micro-structures preserved within the samples, including crystalline dendrites and micron-scale iron-rich spheres with liquid immiscibility textures. We assert these micro-structures reflect local redox conditions and cooling rates and can inform interpretation of high temperature events, enabling new insights into fireball condensation physics and chemistry when metals from the local environment (i.e. structural steel) are vaporized or entrained. Finally, these observations also significantly expand the availability of silicate immiscibility datasets applicable to rapidly quenched systems such as meteorite impact melt glass.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2021

Electrochemical study of the dissolution of oxide films grown on type 316L stainless steel in molten fluoride salt

The corrosion behavior of oxide films grown on Type 316L stainless steel (SS) in molten FLiNaK (LiF-NaF-KF: 46.5−11.5−42 mol.%) salt was investigated. The results show that the oxide film formed on Type 316L SS is unstable and can only temporarily protect materials from corrosion in molten FLiNaK salt. Based on the electrochemical impedance spectroscopy, the oxide dissolution rate is calculated to be 0.85 nm/h at 700 °C in molten FLiNaK salt. Finally, after the oxide film dissolved, Cr and Fe are selective dealloyed from the steel, leading to intergranular corrosion of Type 316L SS in molten fluoride salt.

EIS↗
DOE OSTI2021

Bulk and Short-Circuit Anion Diffusion in Epitaxial Fe 2 O 3 Films Quantified Using Buried Isotopic Tracer Layers

Self-diffusion is a fundamental physical process that, in solid materials, is intimately correlated with both microstructure and functional properties. In this work, a universal approach is presented to precisely characterize self-diffusion in ionic solids by isotopically enriching anions and/or cations at specific locations within an epitaxial film stack, and characterize their redistribution at high spatial resolution with atom probe tomography. Nanoscale anion diffusivity is quantified in epitaxial α-Fe 2 O 3 thin films deposited by molecular beam epitaxy with a thin (10 nm) buried layer highly enriched in 18 O. The isotopic sensitivity of the atom probe allows precise measurement of 18 O distribution across the sharp interfaces between this layer and the surrounding Fe 2 O 3 after annealing. Short-circuit anion diffusion through 1D and 2D structural defects in Fe 2 O 3 are also directly visualized in 3D. This versatile approach to study precisely tailored thin film samples at high spatial and mass fidelity will facilitate a deeper understanding of atomic-scale diffusion phenomena.

36 MATERIALS SCIENCE↗
DOE OSTI2021

High temperature fracture toughness of single-layer CrAlN and CrAlSiN hard coatings

Hardness and toughness are among of the most important physical properties of hard coatings and it is important to analyze them at temperatures present during coating applications. CrAlN and CrAlSiN are hard coating materials frequently used in high temperature applications. Here, while both possess high oxidation resistance and thermal stability, they differ in their microstructure. In this work, the fracture toughness of magnetron-sputtered single-layer CrAlN and CrAlSiN hard coatings was determined up to 700 °C by in-situ scanning electron microscopy micro-cantilever deflection tests. The results show a decreasing trend of the fracture toughness with increasing temperature. While both coatings' fracture toughness trends are similar up to 400 °C, they diverge at higher temperatures. In general, the evolution of the fracture toughness with increasing temperature shows similarities to the observed changes in hardness in the same temperature range.

36 MATERIALS SCIENCE↗